[Federal Register Volume 80, Number 125 (Tuesday, June 30, 2015)]
[Notices]
[Pages 37312-37314]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-16034]



[[Page 37312]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0160]


NuScale Power, LLC, Design-Specific Review Standard and Safety 
Review Matrix

AGENCY: Nuclear Regulatory Commission.

ACTION: Design-specific review standard; request for comment.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is soliciting 
public comment on the Design-Specific Review Standard (DSRS) and Safety 
Review Matrix for the NuScale Power, LLC, design (NuScale DSRS Scope 
and Safety Review Matrix). The purpose of the NuScale DSRS is to 
provide guidance to NRC staff in performing safety reviews where 
existing NUREG-0800, ``Standard Review Plan for the Review of Safety 
Analysis Reports for Nuclear Power Plants: LWR Edition,'' Standard 
Review Plans (SRP) have been modified by the staff specifically for the 
NuScale design, or do not address unique features of the NuScale 
design. The DSRS also allows NRC staff to more fully integrate the use 
of design-specific risk insights into the review of the NuScale design 
certification application (DC) or an early site permit (ESP) or 
combined license (COL) application that references the NuScale design.

DATES: Submit comments by August 31, 2015. Comments received after this 
date will be considered, if it is practical to do so, but the NRC is 
able to ensure consideration only for comments received on or before 
this date.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0160. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Jenny Gallo, Office of New Reactors, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; 
telephone: 301-415-7367; email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0160 when contacting the NRC 
about the availability of information regarding this document. You may 
obtain publicly-available information related to this action by any of 
the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0160.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section. The NuScale DSRS Scope and Safety 
Review Matrix is available in ADAMS under Accession No. ML15156B063.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2015-0160 in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Further Information

A. Background

    In the Staff Requirements Memorandum (SRM) COMGBJ-10-0004/COMGEA-
10-0001, ``Use of Risk Insights to Enhance the Safety Focus of Small 
Modular Reactor Reviews,'' dated August 31, 2010 (ADAMS Accession No. 
ML102510405), the Commission provided direction to the NRC staff on the 
preparation for, and review of, small modular reactor (SMR) 
applications, with a near-term focus on integral pressurized-water 
reactor designs. The Commission directed the NRC staff to more fully 
integrate the use of risk insights into pre-application activities and 
the review of applications and, consistent with regulatory requirements 
and Commission policy statements, to align the review focus and 
resources to risk-significant structures, systems, and components and 
other aspects of the design that contribute most to safety in order to 
enhance the effectiveness and efficiency of the review process. The 
Commission directed the NRC staff to develop a design-specific, risk-
informed review plan for each SMR design to address pre-application and 
application review activities. An important part of this review plan is 
the DSRS. The DSRS for the NuScale design is the result of the 
implementation of the Commission's direction.

B. DSRS for the NuScale Design

    The NuScale DSRS reflects current NRC staff safety review methods 
and practices which integrate risk insights and, where appropriate, 
lessons learned from the NRC's reviews of DC and COL applications 
completed since the last revision of the NUREG-0800, SRP Introduction, 
Part 2, ``Standard Review Plan for the Review of Safety Analysis 
Reports for Nuclear Power Plants: Light-Water Small Modular Reactor 
Edition,'' January 2014 (ADAMS Accession No. ML13207A315). The NuScale 
DSRS Scope and Safety Matrix provides a complete list of SRP sections 
and identifies which SRP sections will be used for DC, COL, or ESP 
reviews concerning the NuScale design; which SRP sections are not 
applicable to the

[[Page 37313]]

NuScale design; and which new DSRS sections are design-specific to 
NuScale. The NuScale DSRS Scope and Safety Review Matrix is available 
in ADAMS under Accession No. ML15156B063.
    The NRC staff is soliciting public comment on the NuScale DSRS 
Scope and Safety Review Matrix and the individual NuScale-specific DSRS 
sections referenced in the table below. Specifically, the NRC requests 
comment on the sufficiency of the scope of the proposed NuScale review, 
as encompassed by the Safety Review Matrix, and on the technical 
content of the individual NuScale-specific DSRS sections identified in 
the table below. These sections were revised from the relative SRP 
sections or developed to incorporate design-specific review guidance 
based on features of the NuScale design. The NRC is not soliciting 
general comments on NUREG-0800 sections that are designated with the 
applicability ``A) Use SRP Section'' in the Safety Review Matrix, but 
specific comments on the adequacy of these NUREG-0800 sections for use 
in the review of the NuScale design certification application will be 
considered.

----------------------------------------------------------------------------------------------------------------
             Section                  Design-specific review standard title           ADAMS  Accession No.
----------------------------------------------------------------------------------------------------------------
Matrix...........................  NuScale Power, LLC DSRS Scope and Safety     ML15156B063
                                    Review Matrix.
3.11.............................  Environmental Qualification of Mechanical    ML15131A247
                                    and Electrical Equipment.
3.13.............................  Threaded Fasteners--ASME Code Class 1, 2,    ML15084A277
                                    and 3.
3.3.1............................  Offsite Power System.......................  ML15071A259
3.3.2............................  Tornado Loads..............................  ML15071A267
3.4.1............................  Internal Flood Protection for Onsite         ML15139A112
                                    Equipment Failures.
3.4.2............................  Analysis Procedures........................  ML15071A324
3.5.1.1..........................  Internally Generated Missiles (Outside       ML15139A081
                                    Containment).
3.5.1.2..........................  Internally Generated Missiles (Inside        ML15139A096
                                    Containment).
3.5.1.3..........................  Turbine Missiles...........................  ML15070A248
3.5.1.4..........................  Missiles Generated by Tornadoes and Extreme  ML15139A121
                                    Winds.
3.5.2............................  Structures, Systems, and Components to be    ML15139A102
                                    Protected from Externally-Generated
                                    Missiles.
3.5.3............................  Barrier Design Procedures..................  ML15071A273
3.7.1............................  Seismic Design Parameters..................  ML15084A279
3.7.2............................  Seismic System Analysis....................  ML15084A177
3.7.3............................  Seismic Subsystem Analysis.................  ML15131A340
3.8.2............................  Steel Containment..........................  ML15131A373
3.8.4............................  Other Seismic Category I Structures........  ML15118A151
3.8.5............................  Foundations................................  ML15132A186
4.2..............................  Fuel System Design.........................  ML15132A517
4.3..............................  Nuclear Design.............................  ML15125A374
4.4..............................  Thermal and Hydraulic Design...............  ML15131A427
4.5.2............................  Reactor Internal and Core Support Structure  ML15070A325
                                    Materials.
4.6..............................  Functional Design of Control Rod Drive       ML15119A111
                                    System.
5.2.2............................  Overpressure Protection....................  ML15118A931
5.2.4............................  Reactor Coolant Pressure Boundary Inservice  ML15125A305
                                    Inspection and Testing.
5.2.5............................  Reactor Coolant Pressure Boundary Leakage    ML15132A194
                                    Detection.
5.3.1............................  Reactor Vessel Materials...................  ML15070A457
5.3.2............................  Pressure-Temperature Limits,                 ML15070A468
                                    Upper[dash]Shelf Energy, and Pressurized
                                    Thermal Shock.
5.3.3............................  Reactor Vessel Integrity...................  ML15070A462
5.4..............................  Rx Coolant System Component and Subsystem    ML15126A156
                                    Design.
5.4.2.1..........................  Steam Generator Materials..................  ML15131A376
5.4.2.2..........................  Steam Generator Program....................  ML15070A562
5.4.7............................  Residual Heat Removal (RHR) System.........  ML15131A360
5-4 BTP..........................  Design Requirements of the RHR System......  ML15132A524
6.1.1............................  Engineered Safety Features Materials.......  ML15070A567
6.1.2............................  Protective Coating Systems (Paints)--        ML15071A372
                                    Organic Materials.
6-1 BTP..........................  pH for Emergency Coolant Water for PWRs....  ML15125A369
6.2.1............................  Containment Functional Design..............  ML15118A922
6.2.1.1.A........................  PWR Dry Containments, Including Sub-         ML15118A264
                                    atmospheric Containments.
6.2.1.3..........................  Mass and Energy Release Analysis for         ML15112A134
                                    Postulated Loss-of-Coolant Accidents
                                    (LOCAs).
6.2.1.4..........................  Mass and Energy Release Analysis for         ML15118A293
                                    Postulated Secondary System Pipe Ruptures.
6.2.2............................  Containment Heat Removal Systems...........  ML15131A341
6.2.4............................  Containment Isolation System...............  ML15119A087
6.2.5............................  Combustible Gas Control in Containment.....  ML15119A090
6.2.6............................  Containment Leakage Testing................  ML15119A084
6.2.7............................  Fracture Prevention of Containment Pressure  ML15112A517
                                    Boundary.
6.3..............................  Emergency Core Cooling System..............  ML15125A322
6.6..............................  Inservice Inspection and Testing of Class 2  ML15127A136
                                    and 3 Components.
7.0..............................  Instrumentation and Controls--Introduction   ML15125A340
                                    and Overview of Review Process.
7.0, A...........................  Instrumentation and Controls--Hazard         ML15132A583
                                    Analysis.
7.0, B...........................  Instrumentation and Controls--System         ML15132A603
                                    Architecture.
7.0, C...........................  Instrumentation and Controls--Simplicity...  ML15132A611
7.0, D...........................  Instrumentation and Controls--References...  ML15132A618
7.1..............................  I&C--Fundamental Design Principles.........  ML15125A335
7.2..............................  Instrumentation and Controls--System         ML15125A360
                                    Characteristics.
8.1..............................  Electric Power--Introduction...............  ML15146A269
8.2..............................  Offsite Power System.......................  ML15125A425
8-2 BTP..........................  Use of Diesel-Generator Sets for Peaking...  Ml15131A386
8.3.1............................  AC Power Systems (Onsite)..................  ML15125A384

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8.3.2............................  DC Power Systems (Onsite)..................  ML15125A386
8-3 BTP..........................  Stability of Offsite Power Systems.........  ML15125A390
8.4..............................  Station Blackout...........................  ML15126A149
8-6 BTP..........................  Adequacy of Station Electric Distribution    ML15131A461
                                    System Voltages.
9.1.2............................  New and Spent Fuel Storage.................  ML15125A307
9.1.3............................  Spent Fuel Pool Cooling and Cleanup System.  ML15146A034
9.2.6............................  Condensate Storage Facilities..............  ML15131A245
9.3.2............................  Process and Post-Accident Sampling Systems.  ML15131A298
9.3.4............................  Chemical and Volume Control System (PWR)     ML15131A305
                                    (Including Boron Recovery System).
9.3.6............................  Containment Evacuation and Flooding Systems  ML15112A190
9.5.2............................  Communications Systems.....................  ML15084A403
9.5.3............................  Lighting Systems...........................  ML15112A148
10.2.............................  Turbine Generator..........................  ML15126A086
10.2.3...........................  Turbine Rotor Integrity....................  ML15127A046
10.3.............................  Main Steam Supply System...................  ML15131A329
10.4.1...........................  Main Condensers............................  ML15127A049
10.4.2...........................  Main Condenser Evacuation System...........  ML15127A349
10.4.3...........................  Turbine Gland Sealing System...............  ML15126A477
10.4.4...........................  Turbine Bypass System......................  ML15131A417
10.4.5...........................  Circulating Water System...................  ML15126A467
10.4.6...........................  Condensate Cleanup System..................  ML15118A943
10.4.7...........................  Condensate and Feedwater System............  ML15126A470
10.4.10..........................  Auxiliary Boiler System....................  ML15131A261
11.1.............................  Source Terms...............................  ML15112A526
11.2.............................  Liquid Waste Management System.............  ML15124A607
11.3.............................  Gaseous Waste Management System............  ML15112A694
11.4.............................  Solid Waste Management System..............  ML15119A057
11.5.............................  Process and Effluent Radiological            ML15118A609
                                    Monitoring Instrumentation and Sampling
                                    Systems.
11.6.............................  Guidance on I&C Design Features for Process  ML15125A367
                                    and Effluent Radiological Monitoring and
                                    Area Radiation and Airborne Radioactivity
                                    Monitoring.
12.2.............................  Radiation Sources..........................  ML15070A194
12.3-12.4........................  Radiation Protection Design Features.......  ML15070A204
12.5.............................  Operational Radiation Protection Program...  ML15070A210
14.2.............................  Initial Plant Test Program--Design           ML15084A407
                                    Certification and New License Applicants.
14.3.2...........................  Structural and Systems Engineering--         ML15084A411
                                    Inspections, Tests, Analyses, and
                                    Acceptance Criteria.
14.3.4...........................  Reactor Systems--Inspections, Tests,         ML15125A294
                                    Analyses, and Acceptance Criteria.
14.3.5...........................  Instrumentation and Controls--Inspections,   ML15127A383
                                    Tests, Analyses, and Acceptance Criteria.
14.3.6...........................  Electrical Systems--Inspections, Tests,      ML15127A373
                                    Analyses, and Acceptance Criteria.
14.3.7...........................  Plant Systems--Inspections, Tests,           ML15131A328
                                    Analyses, and Acceptance Criteria.
15.0.............................  Introduction--Transient and Accident         ML15125A297
                                    Analyses.
15.0.3...........................  Design Basis Accidents Radiological          ML15127A387
                                    Consequence Analyses for Advanced Light
                                    Water Reactors.
15.1.1-15.1.4....................  Decrease in FW Temperature, Increase in FW   ML15127A391
                                    Flow, Increase in Steam Flow, and
                                    Inadvertent Opening of a Steam Generator
                                    Relief or Safety Valve.
15.1.5...........................  Steam System Piping Failures Inside and      ML15125A317
                                    Outside of Containment (PWR).
15.1.6...........................  Loss of Containment Vacuum.................  ML15127A395
15.2.1-15.2.5....................  Loss of External Load; Turbine Trip; Loss    ML15127A400
                                    of Condenser Vacuum; Closure of Main Steam
                                    Isolation Valve (BWR); and Steam Pressure
                                    Regulator Failure (Closed).
15.2.6...........................  Loss of Non-Emergency AC Power to the        ML15125A292
                                    Station Auxiliaries.
15.2.7...........................  Loss of Normal Feedwater Flow..............  ML15125A293
15.2.8...........................  Feedwater System Pipe Breaks Inside and      ML15118A927
                                    Outside Containment (PWR).
15.4.1...........................  Uncontrolled Control Rod Assembly            ML15118A482
                                    Withdrawal from a Subcritical or Low Power
                                    Startup Condition.
15.4.2...........................  Uncontrolled Control Rod Assembly            ML15118A600
                                    Withdrawal at Power.
15.4.3...........................  Control Rod Misoperation (System             ML15131A364
                                    Malfunction or Operator Error).
15.4.6...........................  Inadvertent Decrease in Boron Concentration  ML15118A474
                                    in the Reactor Coolant (PWR).
15.5.1-15.5.2....................  Chemical and Volume Control System           ML15125A463
                                    Malfunction that Increases Reactor Coolant
                                    Inventory.
15.6.5...........................  LOCAs Resulting From Spectrum of Postulated  ML15131A334
                                    Piping Breaks Within the Reactor Coolant
                                    Pressure Boundary.
15.6.6...........................  Inadvertent Opening of a PWR Pressurizer     ML15125A467
                                    Pressure Relief Valve.
15.9A............................  Thermal-hydraulic Stability................  ML15131A311
16.0.............................  Technical Specifications...................  ML15131A316
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    Dated at Rockville, Maryland, this 23rd day of June 2015.

    For the Nuclear Regulatory Commission.
Jenny M. Gallo,
Project Manager, Small Modular Reactor Licensing Branch, Division of 
Advanced Reactors and Rulemaking, Office of New Reactors.
[FR Doc. 2015-16034 Filed 6-29-15; 8:45 am]
 BILLING CODE 7590-01-P